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- 701 - STUDIES ON THE PREPARATION AND CHARACTERISATION OF AMMONIUM URANYL CARBONATE <AUC> V.N. Krishnan, M.S. Visweswaraiah, P.D. Shr i ngarpure. K.S. Koppiker Uranium Extraction Division, B.A.R.C. V.G. Date Atomic Fuels Division, B.A.R.C. Studies have been carried out in the laboratory on the preparation of Ammonium uranyl carbonate (AUO, using concentrated .solution o-f uranyl nitrate. The precipitation o-f AUC has been done by the addition o-f ammonium carbonate solution and secondly by injecting gaseous ammonia and carbon dioxide. The precipitates obtained under varying parameters have been characterised by chemical and XRD analysis and the precipitates obtained under ideal conditions have been -found to have the -formula C<NH4>4U02 <C03>3J, Though, the studies were mainly aimed at standardising the procedure -for the preparation and i dent i-f icat i on o-f AUC powders, some of the powders have been tested, to see their suitability -for conversion to ceramic grade UO2 powder and its pel 1etisation and sintering properties o-f " the gellets. The data collected during these studies is presented. INTRODUCTION Uranium dioxide <U02> has established itself as one of the most successful fuels in nuclear power reactors, essentially water cooled type- Production of UO2 powder has been reported both by wet and dry processes. In the wet process ammonium diuranate (ADU) and ammonium uranyl carbonate <AUC> are precipitated from uranyl nitrate solution by ammonia / ammonium carbonate solutions. In the dry process UF^ is decomposed, reduced by steam and hydrogen in fluidised bed/ rotary kilns to get U02 powder. Nukem, in West Germany, has been in the forefront in developing the processes to convert UF^ / uranyj nitrate to UO2 with AUC as an intermediate product <1,2>. Several recent publications also give the improvements made in the AUC process <3-15). AUC is produced by the combination of uranyl nitrate and ammonium carbonate at specific pH and temperature. It is also prepared by passing ammonia and carbon dioxide gases into uranyl nitrate solution.. In an - 702 extractive process,the organic extract of uranium is treated with ammonium carbonate solution (11). In these processes ammonium carbonate solution / ammonia and carbon dioxide gases used were several times the stoichiometric requirement but the powders obtained are crystalline and -free -flowing. The -following are the reactions for the liquid and gaseous precipitations respectively. +3(NH CCI ::<NH ) U0 <C0 ) +2NH N0 U02(N03)2 4>2 3 4 4 2 3 3 <» 3 + U02(N03>2 6NH3+3C02+3H20=<NH4>4U02<C03)3+2NH4N03 Yi Ming Pan et al. <5> treated uranyl nitrate solution <200-400 g. U/l ) with ammonia and ammonium carbonate under various experimental conditions. The recommended' procedure was to use a mixture o-f ammonia (10X w/v) and ammonium carbonate ( 25X w/v) solutions at 43 C and a pH of ~ 8.5.The mixture was kept under constant agitation. Ammonia solution used was that required to raise the pH o-f the reaction solution to 7 and the ammonium carbaonate to give a C/U ratio o-f 7.5. I.S. Chang et al. <13> have reported the preparation of AUC using nuclear grade uranyl nitrate solution together with ammonia and carbon dioxide gases. Similar work has also been reported by N. Swaminathan et al . (14) and A. Boulia and A. Mel 1 ah (15). In the present work, an attempt was made to prepare AUC by following two routes,using ammonia / ammonium carbonate solutons and secondly gaseous additions of ammonia and carbon dioxide to uranyl nitrate solution. The powders obtained were characterised by chemical and XRD analyses and measurement of surface area and particle size. A few samples were also converted to U02 to study their pel 1etisation and sintering properties. ~ EXPERIMENTAL A). Ammonia and ammonium carbonate solutions: To uranyl nitrate solution (50 / 200 g U/l), maintained at 40 C, a mixture of ammonia <1 OX w/v) and ammonium carbonate (25Xw/v) solutions were added with constant agitation. The amount of ammonia and ammonium carbonate were fixed as that required to raise the pH of the solution to 7 and to maintain a C/U ratio of '7.5 respectively. When the addition was complete, the already precipitaed ADU dissolved and slowly the AUC started precipitating out. The slurry was allowed to stand overnight for completion of prec.i p i t at i on . The precipitate was then filtered, washed with ethenol and dried at 80 C. The powder was then analysed and the results are shown in Table I. B). Gaseous precipitation! The experiments, mentioned earlier, were repeated by injecting ammonia and carbon - 703 - dioxide gases into the uranyl nitrate solution under agitation, through a single nozzle, with and without diluting the ammonia gas with air. The temperature of the solution was maintained at 60 C and -final pH at S.5 to 9.0. After reaching this pH, the rate of ammonia gas addition was reduced to 50 7. while the carbon dioxide and air addition rates remained unchanged and the addition o-f gases continued -for a further period of 4-5 hours. The precipitate was then filtered, washed and dried as mentioned in the earlier experiment. The dried powders were then tested and the results are shown in Table II. RESULTS AND DISCUSSION. In the case of experiments where the liquid precipitant was used to precipitate AUC from dilute uranium solutions <50 g/1) , the recovery was only about 55X and it was about 85X from concentrated uranium solutions <200 g/1> . However the recovery was over ??X in the case of gaseous precipitation from concentrated uranium solutions. AUC powders obtained from all the above mentioned experiments showed near stoichiometric chemical. composition. X-ray diffraction (XRD) analysis confirmed the standard pattern for the formula C (NH^) ^02(003) 3] mentioned in the ASTM card 27-1018 in all the cases except where ammonia was not diluted with air. In the later case the crystals obtained were needle like with length to breadth ratio of about 7 to 10 unl ike in the ?arl ier cases where this ratio was only 2 to 3 (see figures 1,2 &3>. Surface area of the powders obtained by gaseous precipitation showed a tenfold increase compared to the powders prepared using solution precipitant. The AUC powders confirming to chemical stoichiometry and adherence to the standard ASTM pattern have, on conversion to UO2, exhibited good compacting and sintering behaviour with no need for grannulation and precompacting. They also gave pellets with green densities 5.6 to 6.3 g/cc and a sintered density of 10.24 to 10.63 g/cc. CONCLUSION The recommended parameters for the preparation of AUC have been found to give powders with the required properties which on conversion to UO2, may reasonably be expected to give oxide powder with good pel 1 etisation and sintering behaviour. However, a few large scale trials will have to be undertaken before the findings of this work is to be put to pract i ce. - 704 - ACKNOWLEDGEMENT The authors would like to extend their thanks to Dr.V.S. Jakkal o-f Water Chemistry Division and Shri. A.R. Biswas o-f Atomic Fuels Division, -for their help in obtaining XRD analysis and SEM micrographs. REFERENCES 1. F. Ploger and H.Vietzke, Chemie. Ing. Techn., 37, 7, <1965> 692. 2. K.G. Hackstein and F. Ploger, Atomwirstsch. Atomtechn. 12, (1967) 524. 3. J. Sondermann, J. Nucl. Mat., 106, <1982> 45. <4. M. Becker, U.S. Patent 3963828 U976). 5; Yi- Ming Pan, Che- Bas Ma and Nien-Nan Hsu, J. Nucl. Mat., 99 (1981) 135. 6. I.S. Chang et al. Annual Report KAERI/ RR 240/20, 1980. 7. J.H. Park et al. " 454/84, 1984. 8. J.H. Park et al. " 511/85, 1986. 9. C.S. Choi et al. J. Nucl. Mat. 153 <1988) 148. 10. V. Baron, Atomic Energy Review, 17 <1979> 891. 11. K>. Baron et al . Iriorg. Chem. Acta. 81 <1984> 83. 12. M.S. Visweswaraiah and P.D. Inamdar and K. Subramanian, Symposium on Sintering and Sintering Products, B.A.R.C. Bombay, Oct. 29-31, 1979. 13. I.S. Chang, S.T. Hwang and J.H. park, * • I.A.E.A., Vienna, 7-10 April, 1986. p 63. 14. N. Swaminathan et a!., Advances in Uranium Re-fin ing and Conversion, Proceedings o-f A Technical Committee Meeting on Advances in Uranium Re-fining and Conversion. I.A.E.A., Vienna, 7- 10 April, 1986. P 21. 15. A. Boulia and A. Mel 1 ah, Hydrometallurgy, 21 <1989> 331. - 705 - TABLE : 1 EXPERIMENTAL PARAMETERS AND RESULTS FOF AUC PRODUCTION BY SOLUTION ADDITION Uriniin concentration: Final pH. Amonia solution » Amoniun carb. solution C/U = 7.5 (200 o/l) (8.5-9.0) (10 '/. vM (25 '/. «/v> AUC CHEMICAL ANALYSIS AUC PHYSICAL ANALYSIS X-RAY D1FFRN. SINTERED DENSITY C00E AS PER OF •/. U02 ZNH4 7. C03 S.AREA P.SIZE DENSITY TAP. DEN. ASTN - 27/1018 PELLETS M/6) (111) (6./CC) •(B./CC) (6./CC.) AUC-13. 48.12 14.8? 31.55 0.24 19.8 2.94 1.42 HATCHES 10.41 - 10.49 AUC-14. 51.54 14.41 32.04 0.18 21.0 2.97 1.43 HATCHES 10.37 - 10.41 AUC-15. 51.84 17.53 29.97 0.19 25.0 3.04 1.18 HATCHES NOT DONE AUC-14. 51.78 14.84 29.44 0.31 23.0 2.84 1.13 HATCHES 10.24 - 10.35 AUC-17.
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  • Analysis of Mo Production Capacity in Uranyl Nitrate Aqueous

    Analysis of Mo Production Capacity in Uranyl Nitrate Aqueous

    A. Isnaeni,Atom atIndonesia al., / Atom Vol. Indonesia 40 No. 1 Vol.(2014) 40 40No. - 143 (2014) Analysis of 99Mo Production Capacity in Uranyl Nitrate Aqueous Homogeneous Reactor using ORIGEN and MCNP A. Isnaeni*, M.S. Aljohani, T.G. Aboalfaraj and S.I. Bhuiyan Nuclear Engineering Department, King Abdulaziz University P.O. Box 80240 Jeddah, Kingdom of Saudi Arabia A R T I C L E I N F O A B S T R A C T Article history: 99mTc is a very useful radioisotope in medical diagnostic procedure. 99mTc is Received 11 January 2014 produced from 99Mo decay. Currently, most of 99Mo is produced by irradiating 235U Received in revised form 18 February 2014 in the nuclear reactor. 99Mo mostly results from the fission reaction of 235U targets Accepted 28 February 2014 with a fission yield about 6.1%. A small additional amount is created from 98Mo 99 neutron activation. Actually Mo is also created in the reactor fuel, but usually we Keywords: 99 do not extract it. The fuel will become spent fuel which is a highly radioactive Mo waste. 99Mo production system in the aqueous homogeneous reactor offers a better Uranyl nitrate method, because all of the 99Mo can be extracted from the fuel solution. Fresh Homogeneous reactor reactor fuel solution consists of uranyl nitrate dissolved in water. There is no MCNP separation of target and fuel in an aqueous homogeneous reactor where target and ORIGEN fuel become one liquid solution, and there is no spent fuel generated from this reactor. Simulation of the extraction process is performed while reactor in operation (without reactor shutdown).